The following relates to the nuclear reactor arts, nuclear reactor operating arts, nuclear reactor safety arts, and related arts.
The nuclear island of a nuclear power plant includes a nuclear reactor and a steam generator housed inside a containment structure (sometimes simply called containment), along with various auxiliary systems. The containment is typically a steel or steel-reinforced concrete structure designed to contain any radioactive emissions.
The nuclear reactor is typically of the boiling water reactor (BWR) variety or the pressurized water reactor (PWR) variety. In BWR designs the steam generator is omitted as radioactive steam generated by water boiling inside the BWR directly drives the electrical power generating turbine. PWR designs generate subcooled water. The subcooled water heats feedwater in a steam generator to generate the non-radioactive working steam that drives the turbine. The steam generator is typically located outside the pressure vessel (but still inside containment) and is connected with the reactor by a primary coolant loop of large-diameter piping. However, in integral PWR designs the steam generator is a component housed inside the pressure vessel. In either PWR design, the steam generator serves as a heat sink for the nuclear reactor.
Auxiliary non-safety systems include the pressurizer and a reactor coolant inventory and purification system (RCI). In PWR designs, the pressurizer contains a steam bubble whose pressure can be increased by heating (e.g., with resistive heaters) or decreased by cooling (e.g. by sparging cool water or steam into the steam bubble). The pressurizer communicates with the pressure vessel through a baffle plate (in the case of an integral pressurizer) or via piping (in the case of an external pressurizer), and therefore provides buffered control of primary coolant pressure inside the pressure vessel. The RCI maintains the primary coolant water level in the pressure vessel during normal reactor operation by performing “let down” to remove coolant from the pressure vessel, or injecting make-up water into the pressure vessel. The RCI also maintains an inventory of purified water outside of the pressure vessel for use as make-up water. The nuclear reactor may also include reactor coolant pumps (RCPs) to assist or drive primary coolant circulation in the pressure vessel. Alternatively, natural circulation driven by the hot reactor core may suffice.
A control rods system includes control rods comprising neutron poison that are inserted into guide tubes passing through the reactor core. Controlled partial rod insertion (i.e. “gray rod” operation) enables precise control of the nuclear chain reaction. On the other hand, rapid full insertion of the rods (i.e., SCRAM) immediately shuts down the nuclear chain reaction. (However, unstable intermediate reaction products continue to generate decay heat long after the chain reaction is extinguished). Control rod drive mechanisms (CRDMs) including motors operate the control rods. A given control rod drive can have gray rod functionality, shutdown functionality, or both. The CRDMs are typically located outside of the pressure vessel, conventionally below the vessel in BWR designs and above the vessel in PWR designs. However, integral CRDM designs are known in which the CRDMs are located inside the pressure vessel.
Safety systems include an emergency core cooling system (ECC) that provides high pressure decay heat removal from the pressure vessel to an in-containment heat reservoir such as a refueling water storage tank (RWST) located inside containment, and also includes a high pressure water injection system for injecting water from the RWST (or another source located inside containment) into the pressure vessel. The ECC may include a borated water tank containing a solution of soluble boron dissolved in water for injection under high pressure into the pressure vessel. Boron is a neutron poison, such that injection of borated water helps terminate the nuclear chain reaction. The safety systems also typically include a mechanism for flooding containment with water, for example sourced from the RWST. The safety systems still further include an ultimate heat sink (UHS) located outside containment into which heat is expelled. The UHS may, for example, be a lake or other large body of water, a cooling tower, or so forth. The purpose of the safety systems is to contain and condense any steam generated by a LOCA or other safety event so as to depressurize the pressure vessel and containment. The condensing produces heat that is rejected to the UHS located outside of containment, for example via a heat exchanger. There is redundancy built-in for all safety systems. For example, the United States Nuclear Regulatory Commission (NRC) requires at least two independent systems for performing each safety operation.
During normal operation, the non-safety systems are operative to maintain the nuclear reactor within a normal operational envelope, e.g. within acceptable pressure and water level ranges. Temperature control is provided by controlling the nuclear chain reaction using the gray rods. Pressure control is provided by the pressurizer. Water level is controlled by the RCI. The temperature and pressure (and, to a lesser extent, the water level) are interrelated.
The non-safety systems are also operative during normal startup and shutdown of the nuclear reactor. Shutdown entails providing orderly termination of the nuclear chain reaction and dissipating residual decay heat until the reactor core cools sufficiently to open the reactor pressure vessel. In one approach, the control rods are inserted to terminate the chain reaction. A low level of residual decay heat continues to be emitted by the reactor core due to spontaneous decay of unstable intermediate reaction products having short half-lives of order minutes to weeks. Since this residual heat is much less than the thermal output of the core during normal operation, the vessel pressure can be lowered, and low pressure decay heat removal systems of the RCI, e.g. a low pressure condenser, can be brought online to dissipate the decay heat. Once at a safe residual thermal output level, the water level can be reduced, again using the RCI, and the pressure vessel safely opened.
In a safety event such as a loss of heat sink (e.g. loss of feedwater to the steam generator, or failure of the turbine), electrical blackout (which can lead to shutdown of the RCPs and other components), or a loss of coolant accident (LOCA), the safety systems are invoked to perform a rapid controlled depressurization and shutdown of the reactor. The safety systems deploy responsive to the reactor going outside of its safe operational envelope, or in response to a specific fault trigger signal (e.g., a turbine trip or RCP trip), and are designed to operate passively (for example, in a PWR the shutdown rods are actively held out of the reactor core and passively fall into the core in response to loss of power) or are powered by standalone batteries or diesel generators. The response typically includes dropping the shutdown control rods (SCRAM) and bringing the ECC online to depressurize the reactor and dissipate the residual decay heat. After reactor shutdown in response to a safety event, the process of bringing the nuclear reactor back online is lengthy. For example, the boron-containing solute injected into the pressure vessel by the ECC must be filtered out of the primary coolant. Water in the UHS must be replenished, and water in the RWST must be entirely replaced (as the ECC injects radioactive steam into the RWST). In addition to such recovery operations, government regulations typically dictate that an analysis of the safety event be completed before authorizing bringing the reactor back online.